• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.3
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• pp.313-331
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• 2022

Natural or native abiotic molecular hydrogen (H₂) is a major component in natural gas, however yet its importance in the global energy sector's usage as clean and renewable energy is underestimated. Here we review the occurrence and geological settings of native hydrogen to demonstrate the much widesprease H₂ occurrence in nature by comparison with previous estimations. Three main types of source rocks have been identified: (1) ultramafic rocks; (2) cratons comprising iron (Fe²⁺)-rich rocks; and (3) uranium-rich rocks. The rocks are closely associated with Precambrian crystalline basement and serpentinized ultramafic rocks from ophiolite and peridotite either at mid-ocean ridges or within continental margin(Zgonnik, 2020). Inorganic geological processes producing H₂ in the source rocks include (a) the reduction of water during the oxidation of Fe²⁺ in minerals (e.g., olivine), (b) water splitting due to radioactive decay, (c) degassing of magma at low pressure, and (d) the reaction of water with surface radicals during mechanical breaking (e.g., fault) of silicate rocks. Native hydrogen are found as a free gas (51%), fluid inclusions in various rock types (29%), and dissolved gas in underground water (20%) (Zgonnik, 2020). Although research on H₂ has not yet been carried out in Korea, the potential H₂ reservoirs in the Gyeongsang Basin are highly probable based on geological and geochemical characteristics including occurrence of ultramafic rocks, inter-bedded basaltic layers and iron-copper deposits within thick sedimentary basin and igneous activities at an active continental margin during the Permian-Paleogene. The native hydrogen is expected to be clean and renewable energy source in the near future. Therefore it is clear that the origin and exploration of the native hydrogen, not yet been revealed by an integrated studies of rock-fluid interaction studies, are a field of special interest, regardless of the presence of economic native hydrogen reservoirs in Korea.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.45-49
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• 2010

Purpose: Several radiopharmaceuticals were used for cisternography. But recently, due to more short acquisition time, high resolution than other radiopharmaceuticals like In-111 DTPA, we were using Tc-99m DTPA in cisternography. Using of Tc-99m DTPA for intrathecal, was not officially recognised by the FDA. And there are matters of aseptic meningitis, muscular tetany, seizures by inappropriate radiopharmaceuticals handling. So, it is necessary to prevent adverse reactions while handling the radiopharmaceuticals using in cisternography. Therefore, this study aims to evaluation of usefulness and procedures for safety of radiopharmaceuticals in cisternography. Materials and Methods: Subjects were 12radioactive tracer vials using in cisternography in 2008 Dec. 16 - 2009 Dec. 30. (1) Radioactive tracer Vial test - We were measured NaPertechnetate radiation dose and volume, normal saline volume for dilution, source volume and dose activity for patient injection. And then, calculated mass of pure DTPA. (2) Bacterial endotoxin test - We performed pyrogen test using by negative/positive control vials which was added normal saline 0.2 mL and added normal saline 0.1 mL, Tc-99m DTPA 0.1 mL in test control vial. And then, reacted by digital hot plate in $37.5^{\circ}C$ for 1 hour and compared of gel-clot in each control vials. (3) Compliance safety procedure - We were checked safety issues and wrote out a safety procedure exam sheet. Results: (1) Radioactive tracer Vial test - Mass of DTPA per dose for patient injection (mg) was 0.88 (mg) on average, and Mass of DTPA per volume for patient injection (mg) was 0.74 (mg) on average. (2) Bacterial endotoxin test - All control test vials showed negative reactions. (3) Compliance safety procedure - We were checked safety issues and wrote out a safety exam sheet in all the exams. So, there were no adverse reactions. Conclusion: We could examine easier to safety tests using by Techscan - DTPA (Mallinckrodt): CaNa3. Each test results were passed the safety tests and there are no adverse reactions. The use of Tc-99m DTPA for cisternography, always has been become an issue. Since it has occur adverse reaction while examine the cisternography using by Tc-99m DTPA, it needs to set up the 'Standard Operating procedures'.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.2
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• pp.101-109
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• 2022

The mobility and transport of radioactive cesium are crucial factors to consider for the safety assessment of high-level radioactive waste disposal sites in granite. The retardation of radionuclides in the fractured crystalline rock is mainly controlled by the hydrochemical condition of groundwater and surface reactions with minerals present in the fractures. This paper reports the experimental results of cesium sorption to the Wonju Granite, a typical Mesozoic granite in Korea, performed in an anaerobic chamber that mimics the anoxic environment of a deep disposal site. We measured the rates and amounts of cesium (¹³³Cs) removed by crushed granite samples in different electrolyte (NaCl, KCl, and CaCl₂) solutions and a synthetic groundwater solution, with variations in the initial cesium concentration (10^-5, 5×10^-6, 10^-6, 5×10^-7 M). The cesium sorption kinetic and isotherm data were successfully simulated by the pseudo-second-order kinetic model (r²= 0.99) and the Freundlich isotherm model (r²= 0.99), respectively. The sorption distribution coefficient of granite increased almost linearly with increasing biotite content in granite samples, indicating that biotite is an effective cesium scavenger. The cesium removal was minimal in KCl solution compared to that in NaCl or CaCl₂ solution, regardless of the ionic strength and initial cesium concentration that we examined, showing that K⁺ is the most competitive ion against cesium in sorption to granite. Because it is the main source mineral of K⁺ in fracture fluids, biotite may also hinder the sorption of cesium, which warrants further research.

• The Korean Journal of Nuclear Medicine
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• v.22 no.1
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• pp.83-92
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• 1988

The cocktails of two $^{131}I$ labeled Monoclonal antibody (MCAB) (Anti CA 19-9 F$(ab')_2$ + Anti CEA $F(ab')_2$ fragment), which react specially, with human gastrointestinal cancers, were administered to 10 patients with colorectal (7), stomach(2) and pancreas(1) cancer for scintigraphic detection. All patients were known or postoperatively recurrent cases, and serum tumor markers, CA 19-9 and CEA, were measured with immunoradiometric assay, just before immunoscintigraphy (ISG). The tumor marker's level in serum is not correlated with positive tumor uptake in ISG. The sensitivity and specificity of ISG in detection of 21 tumor sites, based on surgery, CT, ultrasonography and pathology, were 90.5% and 100% One case of colon cancer showed gall bladder metastasis, which was neglected on CT study. Tumor/non tumor uptake ratio of radiolabelled antibody were progressively increased from day 3 to day 7 during study. We summerized as follows 1) The use of cocktails of CEA and CA 19-9 MCAB $F(at')_2$ increased sensitivity and specificity in ISG. 2) Delayed imaging (later than 5 days) increases sensitivitv and specificity due to exclusion of nonspecific iodine accumulation in stomach and lung. 3) Second tracer technique is essential for anatomical landmark by use of a double isotope scan, but subtraction technique, a possible source of artifacts, is no longer necessory when delayed imaging is performed. 4) It may be possible to use two MCAB cocktails of CA 19-9 and CEA in Radioimmunodetection of stomach and pancreas cancer. In conclusion, ISG using MCAB cocktails, $F(ab')_2$ fragment of anti CA 19-9 and Anti CEA, provide additional opportunity for tumor localization and detection of colorectal and other G-I cancer, such as stomach and pancreas.

• The Journal of the Korea Contents Association
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• v.14 no.2
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• pp.467-474
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• 2014

The relative dose calculated by MCNPX and the relative dose measured by ionization chamber and solid phantoms evaluated the accuracy comparing with Monte Carlo simulation. In order to apply Monte Carlo simulation the intraluminal brachytherapy of extrahepatic bile duct cancer, 192Ir sealed radioactive source replicate, Bile duct and surrounding organs were made using KMIRD phantom based on a South Korea standard man. To check the absorbed dose of normal organs around bile duct, we set the specific effective energy and initial radioactivity to 1 Ci using MCNPX. Evaluation of the accuracy of the Monte Carlo simulation, the difference of the relative dose is the most 1.96% that satisfy the criteria that is the relative error less than 2% suggested by MCNPX code. In addition, The specific effective energy and absorbed dose of normal organs that were relatively adjacent to bile duct such as right side of kidney, liver, pancreas, transverse colon, spinal cord, stomach and small intestine were relatively high. on the contrary, the organs that were relatively distant to bile duct such as left side of kidney, spleen, ascending colon, descending colon and sigmoid colon were relatively low.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.331-337
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• 2018

The decommissioning of nuclear power plants is generally executed in five steps, including preparation, decontamination, cutting/demolition, waste disposal and environmental restoration. So, for efficient decommissioning of nuclear power plants, worker safety, effects compared to cost, minimization of waste, possibility of reuse, etc., shall be considered. Worker safety and measurement technology shall be secured to exert optimal efficiency of nuclear power plant decommissioning work, for which accurate measurement technology for systems and devices is necessary. Typical In-Situ methods for decommissioning of nuclear plants are CZT, Gamma Camera and ISOCS. This study used ISOCS, which can be applied during the decommissioning of a nuclear power plant site without collecting representative samples, to take measurements of the S/G Water Chamber. To validate the measurement values, Microshield and the GEANT4 code was used as the actual method were used for modeling, respectively. The comparison showed a difference of $1.0{\times}10^1Bq$, which indicates that it will be possible to reduce errors due to the influence of radiation in the natural environment and the precision of modeling. Based on the research results of this paper, accuracy and reliability of measurement values will be analyzed and the applicability of the direct measurement method during the decommissioning of NPPs will be assessed.

• Journal of Radiation Protection and Research
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• v.38 no.3
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• pp.132-137
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• 2013

Cryogenic particle detectors have recently been adopted in radiation detection and measurement because of their high energy resolution. Many of these detectors have demonstrated energy resolutions better than the theoretical limit of semiconductor detectors. We report the development of alpha spectrometer using a micro-fabricated magnetic calorimeter coupled to a large-area particle absorber. A piece of gold foil of $2{\times}2{\times}0.05mm^3$ was glued to the paramagnetic temperature sensor made of sputtered Au:Er film to serve as an absorber for incident alpha particles. We performed experiments with 241Am source at cryogen free adiabatic demagnetization refrigerator (CF-ADR). A high energy resolution of 6.8 keV in FWHM was obtained for 5.5 MeV alpha particles.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.107-114
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• 2009

A Compton camera, which is based on Compton kinematics, is a very promising gamma-ray imaging device in that it could overcome the limitations of the conventional gamma-ray imaging devices. In the present study, the image quality of a rotating Compton camera was evaluated by using 4-D Monte Carlo simulation technique and the applicability to nuclear industrial applications was examined. It was found that Compton images were significantly improved when the Compton camera rotates around a gamma-ray source. It was also found that the 3-D imaging capability of a Compton camera could enable us to accurately determine the 3-D location of radioactive contamination in a concrete wall for decommissioning purpose of nuclear facilities. The 4-D Monte Carlo simulation technique, which was applied to the Compton camera fields for the first time, could be also used to model the time-dependent geometry for various applications.